077-05/rs

DCLL Summary of TBM Design and AnalysisSignificant understanding on DCLL has been achieved by the team

General Atomics C. Wong, D. Carosella, M. P. LabarU. of California, Los Angeles M. Abdou, S. Smolentsev,

M. Dagher, S. Sharafat, N. Morley, P. Calderoni, A. Ying, M. Youssef

Consultant, Germany S. MalangU. of Wisconsin, Madison M. Sawan, G. SviatoslavskyORNL P. Fogarty, S. ZinkleIdaho National Laboratory B. MerrillUCSD D. K. SzePNL R. KurtzLLNL S. Reyes

ASSESSMENT AND DESIGN TEAM:

US ITER-TBM Meeting August 10-12, 2005 Idaho Falls, Idaho

0.0 100

5.0 10-9

1.0 10-8

1.5 10-8

2.0 10-8

2.5 10-8

3.0 10-8

0 5 10 15 20 25 30

Tri

tium

Pro

duct

ion

Rat

e (k

g/m

3 .s)

Radial Distance from FW (cm)

Radial Distribution of Tritium Production in LiPb Breeder

Neutron Wall Loading 0.78 MW/m2

DCLL TBM LiPb/He/FS90% Li-6

Front Channel Back Channel

077-05/rs

Outline

• Key features of DCLL concept• Reference reactor blanket• Conceptual TBM module design and analysis completed

Engineering designNeutronicsPbLi MHDHe thermal hydraulicsStructural analysis including disruption analysisTritium extraction and controlSafetyBypass loop

• PIE and hot-cell requirements• Key tasks for preliminary design• Summary

077-05/rs

Key Features of the DCLL Concept

• Cool FW and ferritic steel structure with 8 MPa He (also used for FW/blanket preheat and possible tritium control)

• Breeder is self-cooled Pb-17Li moving at a slow velocity, < 10 cm/s – allowing high Tout (700°C) leading to ηth> 40% (CCGT)

• Use flow channel inserts (FCIs), wherever possible to:– Provide electrical insulation to reduce

MHD pressure drop– Provide thermal insulation to decouple Pb-17Li

bulk flow temperature from wall/structure temperature

– Provide additional corrosion resistance since only nearly stagnant Pb-17Li is in contact with the ferritic steel structural walls

EU design, FED, 61-62, 2002

US ARIES ST

077-05/rs

DCLL Reference Tokamak FW/Blanket Module

First Wall

He Out

Pb-17Li Out

Reference Tokamak

Outboard max. NWL 3 MW/m2Maximum surface heat flux 0.5 MW/m2

Typical 2x3 m Blanket Module

Structural material: RAFS F82HFW/structure coolant: 8 MPa HeSelf-cooled breeder: Pb-17LiGross ηth > 40% with CCGT

Concentric pipes

077-05/rs

DCLL Reference Reactor Design Blanket Module

Cross-section

Exposed view, He flow

Exploded view

Pb-17Li flow

077-05/rs

US ITER-TBM DCLL Pb-17Li Module Initial Design Parameters with 10 cm frame thickness

• ITER scenario: Fusion power–500 MW, burn time-400 s

• Design heat flux: Average at 0.3 MW/m2, max at 0.5 MW/m2

• Design neutron wall loading: 0.78 MW/m2 (Under high poweroperation, the outboard mid-plane could see 1.09 MW/m2)

• Disruption load: 0.55 MJ/m2

• Disruption energy dump during current quench: 0.72 MJ/m2

• Pulse length: 400 s/

077-05/rs

Conceptual Engineering Design

077-05/rs

DCLL ITER-TBM Half Test Port Design

DCLL TBM

1 of 3 ITER Test Ports

077-05/rs

DCLL TBM Mechanical Design (Initial Design)

Test module

Cross-sectionPb-17Li flow

Sections, He flow

Radial thickness of PbLi zone islimited by the volume of PbLi to be

077-05/rs

4041424344454647484950

1 2 3 4 5 6 7 8

OUTLET CHANNELS

VELO

CIT

Y (m

/s)

Header Pass 1 to Pass 2

Header Pass 2 to Pass 3

MINUMUM TARGET VELOCITY42.6 m/s

Header Pass 3 to Pass 4

Header Pass 4 to Pass 5

Helium Gas Flow Distribution Analysis”For the determination of plenum size”

Header Flow Model Flow Distribution Results

Plenum size was increased by ~ 5 cm each until relatively uniform velocities are reached from different outlet channels

old new

077-05/rs

Top Plate Assembly

Bottom plate assembly

Outer back Plate Assembly

He Inlet/Outlet Pipe Assembly

Center divider late Assembly

Pb-Li Inlet/Outlet Pipe Assembly

1940

645

413

DCLL Module W/O FCIRadial depth increased from 308 mm to 413 mm

First wallpanel

077-05/rs

FW & Top/Bottom Plate He Flow Schematic

He Coolant Flow Circuit 1 Thru FW & Back Plate

He Coolant Flow Circuit 2 Thru FW & Back Plate

Top Plate He Flow

077-05/rs

Install the Cross Assembly of the PbLi Dividing Plates and weld to First Wall Module,then install the “Tee” Assembly and finish welding.

( welds are shown in yellow )

“Cross” Assembly“Tee” Assembly

077-05/rs

SiC InsertsFront Upper

Assembly

SiC InsertsRear Assembly

SiC InsertsBottom Assembly

SiC InsertsTop Assembly

Divider / Grid PlateAssembly

Install SiC/SiC Flow Channel Inserts

SiC InsertsFront Lower

Assembly

SiC inserts shownassembled with the

first wall module removed

SiC inserts shownfully assembled infirst wall module

Front View Back View

077-05/rs

First Wall

Grid Plates

Divider Plates

Inner Back PlateStructure

Inner Back PlateCover

Install Inner Back Plate AssemblyMake Internal

Seal Welds(shown in yellow)

077-05/rs

Install Top and Bottom Plate AssembliesTop Plate Cover

Top Plate Structure

BottomPlate Cover

BottomPlate

Structure

Make internal seal weldsto top and bottom of

the grid plates

Make external seal welds to

top and bottom plate assemblies

077-05/rs

Neutronics

077-05/rs

Neutronics for DCLL Reference Reactor Blanket

Blanket thicknessOB 75 cmIB 52.5 cm

Local TBR is 1.328OB contribution 0.995IB contribution 0.333

If neutron coverage for double null divertor is 12% overall TBR will be ~1.17 excluding breeding in divertorregion. Based on 3D results of DC molten salt design we expect blanket to have potential for achieving tritium self-sufficiency

Nuclear energy multiplication 1.136Peak nuclear heating in OB blanket

o FS 36 W/cm3o LL 33 W/cm3o SiC 29 W/cm3

077-05/rs

Tritium Production

Local TBR in the DCLL TBM is only 0.741 because of the small thickness (41.3 cm)

Tritium generation rate in the TBM is 3.2x1017 atom/s (1.59x10-6 g/s) during a D-T pulse with 500 MW fusion power

For a pulse with 400 s flat top preceded by 100 s linear ramp up to full power and followed by 100 s linear ramp down total tritium generation is 7.97x10-4 g/pulse

For the planned 3000 pulses per year the annual tritium production in the TBM is 2.4 g/year

Tritium production in the Be PFC is 2.2x10-9 g/s ⇒ 1.1x10-6 g/pulse ⇒3.3x10-3 g/year

Peak tritium production rate in LiPb is 2.94x10-8 kg/m3s during the D-T pulse

0.0 100

5.0 10-9

1.0 10-8

1.5 10-8

2.0 10-8

2.5 10-8

3.0 10-8

0 5 10 15 20 25 30

Tri

tium

Pro

duct

ion

Rat

e (k

g/m

3 .s)

Radial Distance from FW (cm)

Radial Distribution of Tritium Production in LiPb Breeder

Neutron Wall Loading 0.78 MW/m2

DCLL TBM LiPb/He/FS90% Li-6

Front Channel Back Channel

077-05/rs

Nuclear HeatingPower density in Be PFC 8.6 W/cm3

Peak power density in FS structure 8.2 W/cm3

Peak power density in Pb-17Li 18 W/cm3

Peak power density in SiC FCI 5.9 W/cm3

0

5

10

15

20

0 5 10 15 20 25 30 35 40

FSLiPbSiC

Pow

er D

ensit

y (W

/cm

3 )

Radial Distance from FW (cm)

Radial Distribution of Power Density in DCLL TBM Components

Neutron Wall Loading 0.78 MW/m2

LL

SiC FS

0.033Inlet/Outlet Pipes

0.103Back Wall

0.982Total

0.228Back Pb-17Li Channel

0.395Front Pb-17Li Channel

0.019Flow Channel Divider

0.020Radial Ribs

0.014Top/Bottom Walls

0.042Side Walls

0.128First Wall

Nuclear Heating (MW)

Component

0.014 MWTop/Bottom

0.033 MWPiping

0.020 MWRibs

0.019 MWDivider

0.042 MWSide Walls

0.128 MWFW

0.228 MWBack LL Channel

0.395 MWFront LLChannel

0.103 MWBack

077-05/rs

Radiation Damage in Steel Structure

For 0.57 MW/m2 average NWL and total fluence 0.3 MWa/m2 total irradiated time is 0.526 FPYPeak cumulative end-of-life dpa in FW is 5.7 dpa and He production is 64 He appmPeak cumulative end-of-life He production in inlet/outlet pipes is 0.34 He appm⇒ Pipe connections at back of TBM are reweldable

Determined radial variation of dpa, He and hydrogen production rates in structure

10-1

100

101

102

103

0 5 10 15 20 25 30 35 40

dpa/FPYHe appm/FPYH appm/FPY

Dam

age

Rat

e in

Ste

el S

truc

ture

per

FPY

Depth in Blanket (cm)

Radial Distribution of Damage Rate in Steel Structure

Neutron Wall Loading 0.78 MW/m 2

DCLL TBMLiPb/He/FS

90% Li-6

10-5

10-4

10-3

10-2

10-1

100

101

102

103

0 50 100 150

dpa/FPYHe appm/FPYH appm/FPY

Dam

age

Rat

e in

Ste

el S

truc

ture

per

FPY

Radial Distance from FW (cm)

Radial Distribution of Damage Rate in Steel Structure

Neutron Wall Loading 0.78 MW/m2

DCLL TBMLiPb/He/FS90% Li-6

Pipes Region Shield Plug

DCLLTBM

Dam

age

rate

40 cm 150 cm

077-05/rs

Total Radioactivity and After Heat and Contribution from Each Material

10-5

10-4

10-3

10-2

10-1

100

101

102

103

104

105

106

107

10-1 100 101 102 103 104 105 106 107 108 109 1010 1011

F82HLiPbSiCTotal

Time after Shutdown, Sec

1 s 1 m 1 h 1 d 1 mo 1 y 10 y 100 y 1000 y

TBM dimension64.5 cm Toroidal194 cm poloidal200.2 cm radial (including 128 cm shield)

10-14

10-13

10-12

10-11

10-10

10-9

10-8

10-7

10-6

10-5

10-4

10-3

10-2

10-1

10-1 100 101 102 103 104 105 106 107 108 109 1010 1011

F82HLiPbSiCtotal

Time after Shutdown, Sec

1 s 1 m 1 h 1 d 1 mo 1 y 10 y 100 y 1000 y

TBM dimension64.5 cm Toroidal194 cm poloidal200.2 cm radial (including 128 cm shield)

Radioactivity After Heat

At shutdown, the total radioactivity inventory and after heat are as low as ~2 MCi and 0.02 MW, respectively. They are mainly due to the F82H structure present in the shield, back plate, the back breeder channel, and the FW, in that order

077-05/rs

Waste Disposal Rating

Definition: The radwaste classification of TBM components was evaluated according to the Nuclear Regulatory Commission (NRC) 10FR61 and Fetter waste disposal concentration limits. The limits given are based on the assumption that all solid components are crushed before being disposed (no voids). Components having WDR >1, according to Class C limits, do not qualify for shallow land burial.

Results: The WDR values for F82H structure, the Pb-17Li breeder, and the SiC insert are 1.3x10-2, 8.7x10-3, and 2.1x10-4, respectively, based on Fetter’s limits. Although the Fetter limits are generally more conservative, still the values are much lower than unity and therefore these materials are qualified for shallow land burial according to the Class C limits and are within ITER guidelines.

077-05/rs

PbLi flow including MHD effects

077-05/rs

Effect of Electrical Conductivity and Pressure Equalization hole Geometry of FCI on Velocity Profile and Pressure Drop

With pressure equalization holes

With pressure equalization slot inverse flow generated

ΔP/dx at different σ for PEH and PES

B-fielddirection

077-05/rs

Pb-17Li MHD/Heat Transfer Issues(Pb-17Li, Tin=400° C, Tout=650° C)

1. Effectiveness of SiCf/SiC FCI as electrical/thermal insulator

2. 3-D MHD pressure drop

3. Flow/heat transfer in the concentric pipe

• In the poloidal channel flows, MHD pressure drop is caused by cross-sectional currents, which can be reduced with FCI. At the same time FCI serves as thermal insulator.

• 3-D MHD pressure drop is related to axial currents, which can’t be reduced by insulation. This type of pressure drop is associated with the flows in manifolds and a fringing magnetic field.

• “Hot” Pb-17Li flowing in the internal pipe will cause wall temperature rise. The goal of the analysis is to assess this rise in temperature.

077-05/rs

1. Poloidal Channel Flow with FCI

• 5 mm FCI with σ=20 S/m reduces the MHD pressure drop by a factor of 100

• Near-wall velocity jets exist, indicating that insulation is not perfect

• Velocity in the gap and jet flows will affect heat transfer

Velocity profile in the front poloidalchannel with FCI. Velocity is scaled with the mean velocity; z and y are scaled with b (half of the channel width).

B

077-05/rs

2. Pressure Drop in the Pb-17Li

• The MHD pressure drop is mostly contributed by 3-D flows in the manifolds, where axial currents appear due to changes in the flow direction

• High MHD pressure drop also appears in the flows in the concentric pipe, when the pipe goes through the fringing magnetic field

1000.302-0.442Total

23.30.07-0.14Outlet manifold

23.20.07-0.14Inlet manifold

19.30.0585Concentric pipe (annulus, fringing B-field)

19.30.0585Concentric pipe (internal, fringing B-field)

9.50.0286Concentric pipe (annulus, uniform B-field)

5.115.4×10-3Concentric pipe (internal, uniform B-field)

0.160.485×10-3Return channels

0.13 0.384×10-3Front channels

ΔPi/ΔP(%)

ΔPi(MPa)

Flow

077-05/rs

3. Heat Transfer in the Concentric Pipe

• Velocity profile was calculated first using 2-D MHD code

• Heat transfer calculations were then performed using FLUENT

• Maximum wall temperature does not exceed 475°C

Cross-sectional temperature distribution (°C) in the concentric pipe

B

077-05/rs

First wall design

077-05/rs

Results From First Wall Fluent Code Modeling

• With counter flow temperature variation between adjacent tubes is minimum

• One-sided wall roughening maintains FWtemperature well below 550oC

• One-sided wall roughening increases heat transfer coefficient but increases also pressure drop

C377Pa3716598w/m2-k3455smooth concurrent flow

C373Pa3716598w/m2-k3455smooth concurrent flow

C376Pa5000512w/m2-k9150Ks=0.000395 (one-side)

C376Pa3669589w/m2-k3404Smooth counter flow

He Outlet TP dropmax TFW(C)h

0.16 MW/m2 from PbLiSide

C373Pa5000514w/m2-k9150Ks=0.000395 (one-side)

C373Pa3669589w/m2-k3404Smooth counter flow

He Outlet TP dropmax TFW(C)hAdibataic PbLi Sice

077-05/rs

Counter Flow FW Thermal Performance

Plasma sideHeat Flux =0.3 MW/m2

Pb-17Li sideAdiabatic

First Wall Thermal Model

525 C

430 C First Wall Thermal Results

He Inlet Temp: 360° C

He Outlet Temp: 432° C

Max FW Temp: 523° C

077-05/rs

Power removed by Helium cooling:

0.375 MW

FW surface heat flux contributionAverage heat flux = 0.3 MW / m2FW area = 1.94 m X 0.64 m = 1.25 m2

+ 0.359 MW

Nuclear heating contributionTotal nuclear heat in TBM = 0.982 MWNuclear heat in Pb-17Li = 0.623 MWAssuming no leaks from Pb-17Li because of FCI

= 0.734 MW

Resulting total mass flow rate = 2.353 kg /sInlet T = 360 C / Outlet T = 420 CCp = 5200 J / kg K

(54% of total power)

10.1%0.811 MPaTotal

4.4%0.351 MPaFirst wall

1.7%0.14 MPaFlow distribution

4%0.32 MPaInlet / outlet pipes

Fraction of inlet pressurePressure drop Circuit component

Estimated Total Pressure Drop

DCLL He cooling Energy Balance and Pressure Drops

077-05/rs

Structural Analysis

077-05/rs

Temperature Distribution (Temperature can be controlled via mass flow rate)

(Kelvin)

FW-Helium out at 420oC

FW-Helium inat 360oC

Ribs and Back Plate Helium Tin = 438oC; Tout = 440oC

077-05/rs

FW Thermal Analysis Boundary Conditions

FW Helium makes 5 Passes: Pass 1: In 360oC Out 372oCPass 2: In 372oC Out 384oCPass 3: In 384oC Out 396oCPass 4: In 396oC Out 408oCPass 5: In 408oC Out 420oC

RIB Helium flow:Single PassTin = 420oCTout = 440oC

Heat Transfer Coefficient:FW (plasma side only)hcoef = 6979 W/m2-K*(1-side rib-roughened)All other surfaces

assume smooth wallshcoef = 3586 W/m2-K

5 Pass -Helium flowDirections

Tin = 360oC

Tout = 420oC Heat Loads:Plasma on FWq’’ = 0.5 MW/m2Volumetric Heatingq’’’=8.3 exp (-10 y) MW/m3Leakage from Pb-17Li into walls0.16 MW/m2, total ~ 40% of the heat

Simplified Geometry

077-05/rs

FW – One Sided Roughening Temperature Distribution

{

{Beryllium:Tmax = 580oCF82H:Tmax=559oC(top 1.2mm) h=6978 W/m2K

Higher h or mass flow rate can reduce the Tmax

077-05/rs

Primary + Secondary Stresses

Von Mises(MPa)

077-05/rs

DCLL 5-Channel Section Satisfies ISDC Design Rules

High Temperature Primary Membraneand Membrane Plus Bending Stress Limits are Satisfied

&

m

m

(T ) (T ,t )

mm

t l

SP

S⎧

≤ ⎨⎩

L b mP P K S+ ≤ bL tt

PP SK

+ ≤

Where:K is the bending shape factorand Kt = (K + 1)/2, K=1.15 [1].

1. Tavassoli et al Fusion Engineering and Design 61-62 (2002) 617-628.

077-05/rs

DCLL 5-Channel Section Satisfies ISDC Design Rules

High Temperature Ratcheting Primary Plus Secondary Stress Limits

1X Y+ ≤

Conservative Rule:Cyclic Ratcheting is Prevented

077-05/rs

LOCA Structural Analysis Results

Displacements

Maximum Displacements occur at the tips and is equal to

8.435 mm

Back View

IsoFront View

077-05/rs

Structural Analysis Results

Stress Distribution:

σ = 530 MPa

Critical section 1:(Rib-first wall)

View A

View A

077-05/rs

Reinforced Rib-FW Structure

Adding material (4x4 mm) Reduced Maximum Stresses

from 530 MPa to 450 MPa

σ = 450 MPa

First Wall

Added material

077-05/rs

Reinforced Top Plate Side Wall Contact

Adding material (4x15 mm) Reduced Maximum

Stresses from 616 MPa to 415 MPa

σmax = 415 MPa

077-05/rs

Disruption Eddy Current Analysis

ITER specified cases:18 ms exponential plasma current decay40 ms linear decay… max. JxB force generated

Method of analysis:OPERA E&M analysis code by VectorFieldsA 20° segment of ITERPoloidal field coils, vacuum vessel, TBM are includedPlasma represented by time varying filaments

Simulationvolume

Vacuumvessel

TBM

077-05/rs

• There is a considerable torque about the center of the TBM• There is a large twisting force trying to make a propeller

blade• The grids and the back plate have the greatest forces• Static finite element analysis to determine the stresses is

on-going, and is being iterated with the structural analysis group.

Disruption Analysis Summary“40 ms linear case”

N/mm3-Toroidal Direction (backplate)N/mm3-Radial Direction (backplate)

077-05/rs

Tritium Extraction and High Temperature Loop

077-05/rs

Primary Side Pb-17 Li Vacuum Permeator Scheme

He inletHe outlet

Vacuum pumpVacuum permeator

Blanket Concentric pipes

HeatExchanger

T2 outlet

Pressure boundary (90 °C)

Closed Brayton CyclePb-17Li (460 °C)

Pb-17Li (700 °C)

Pb-17Li pump

Inter-cooler Pre-cooler Recuperator

Turbo-compressor

Power turbine

For Power Reactor

077-05/rs

Single Tube Tritium Permeator Assessment

Mathematical model• A component balance describes the tritium

mass fraction along the membrane length

• Tritium transport to the membrane surface is described by a mass transfer coefficient

• The effective tritium partial pressure at the membrane surface is given by the solubility

• Permeation depends on the permeability

FGMWD

dzdx

ρπ−=

( )0xxMWk

N im

i −=ρ

2

0⎟⎟⎠

⎞⎜⎜⎝

⎛=

sr k

xp

( )prp pplk

AGG −==

_

Vacuum

Pb-17Li in Pb-17Li out

077-05/rs

Tritium Pressure for Various ParametersReasonable Outlet Partial Pressure can be Achieved

Mass transport coefficient 5.6 m/sDifferent tritium breeding ratio (TBR)

Mass transport coefficient 5.6 m/sNiobium permeation reduction factor (PRF)

.025

0 1 2 3 40.0

0.1

0.2

0.3

0.4

0.5

Tube length (m)

Pre

ssur

e (P

a)

.02

.06.025.02

0 1 2 3 4Tube length (m)

0.0

0.1

0.2

0.3

0.4

0.5

Pre

ssur

e (P

a)PRF = 10

PRF = 1

PRF = 50

TBR = 1

TBR = 1.17

Equilibrium results determined by the TMAP code

077-05/rs

While this initial analysis indicates that tritium permeatormay be feasible, many issues remain

• Measured mass transfer coefficients for the Pb-17Li T system• Compatibility of Pb-17Li with Nb at 700°C for power reactor• Additional resistances to tritium permeation such as surface

resistance?• At the Pb-17Li membrane interface, is the effective partial pressure

exerted by tritium indeed given by the solubility equation? (this may be a very different mechanism with a very different rate)

• What pressure can be practically maintained on the permeate side of the membranes?

• Will Nb tubes degrade due to reactions such as oxidation? • Will a surface treatment be needed?• For the DCLL TBM and for demonstration we will replace Nb with

a FS-permeator, and operate external Pb-17Li at T

077-05/rs

TBM Safety

077-05/rs

Reference Accidents Analyzed

1. Ex-vessel LOCA analysis to determine:– Pressurization of TBM vault– Behavior of TBM without active plasma shutdown

2. Coolant leak into TBM breeder zone analysis to assess:– Module and ancillary system pressurization– Chemical reactions and hydrogen formation– Subsequent in-vessel leakage

3. In-vessel TBM coolant leak analysis to demonstrate:– Minimal Pressurization of ITER first confinement barrier (i.e., VV)– Passive removal of TBM decay heat– Limited chemical reactions and hydrogen formation

077-05/rs

MELCOR Developed for TBM Accident Analyses

First wall

Concentric pipe

Permeator

PbLi/He HX

Back plate

He pipes

He/H2O HXs

Vacuum vessel

Be/FS/HE/FS/SiC

Drain tank

Port cell

TWCS vault

• 30 control volumes• 37 flow paths• 72 heat structures

(psuedo 3D TBM conduction)

• 6 valves• 1 rupture disk• 1 pump and 2

circulators

077-05/rs

Ex-vessel LOCA Pressure Results

• LOCA assumed to start at the end of a reactor pulse flat top (300 s)• Pressure port cell relief valve (set to open at 0.2 atm pressure differential with

TWCS vault and to re-seat at 0.01 atm pressure differential) limits test cell pressure to 1.5 atm, not exceeding confinement barrier design limits of 2 atm

0.0 0.5 1.0 1.5 2.0Time (hr)

0.0

0.5

1.0

1.5

2.0

Pre

ssur

e (a

tm)

Test cellTWCS vaultVacuum vessel

2600 2800 3000 3200Time (s)

0.8

1.0

1.2

1.4

1.6

Pre

ssur

e (a

tm)

2.0Test cellTWCS vaultVacuum vessel

Expanded view

077-05/rs

Ex-vessel LOCA Temperature and Oxidation Results

• FW beryllium evaporates and disrupts plasma ~ 90 seconds after LOCA starts• Beryllium on “hot strip” (region with a surface heating of 0.5 MW/m2) reacts with

steam from ITER FW cooling system producing 9 g of H2• In ~100 s, 0.32 m3, of Pb-17Li leaks into the VV; the quantity of hydrogen generated

from PbLi-water reactions based on data* is ~ 2.5x10-4 mol-H2/g LM × 2.92x106 g LM × 2 g-H2/ mol-H2 = 1460 g.

0

250

500

750

1000

1250

1500

Tem

pera

ture

(C)

0.0 0.5 1.0 1.5 2.0Time (hr)

FW

SW 2800 3000 3200Time (s)

0

500

1000

1500

FW

SW

0

1

2

3

FW b

eryl

lium

thic

knes

s (m

m)

FW

.00

.05

.10

.15

FW H

ydro

gen

Prod

uctio

n (k

g)

0.0 0.5 1.0 1.5 2.0Time (hr)

FW hot strip

*D. W. Jeppson, Nuclear Technology/Fusion, 4 (1983), p. 277-287

077-05/rs

TBM Tritium Permeation Analysis

• ITER allowed in-building permeation limit should be ~100 mg-T/a, based on an allowed TBM release to the environment of 1 mg-T as HTO/a and an assumed building detritiation system efficiency of 99%

• A TMAP model was developed to examine permeation from the TBM and ancillary system into the ITER test cell and vault, this model includes:

– All TBM structures (500°C), and the TBM Pb-17Li (temperature change of 400°C to 650°C)

– All piping (400°C or 450°C) and heat exchangers– T2 pulsed production with a flat top value of 1.59e-6 g/s during a 600 s pulse,

which translates into 2.33 g-T/a for 3000 pulses/a– A vacuum permeator, composed of 20, 5 m long, 0.01 m diameter tubes

077-05/rs

TBM Tritium Permeation Results

• TBM tritium concentration reaches an oscillatory equilibrium after ~50 pulses, with Pb-17Li tritium pressure decreasing from 1.5 Pa to 1.1 Pa through permeator by the end of a pulse

• Annual release based on 3000 consecutive pulses will be above ITER in-building annual limit of 100 mg-T/a, so alumina coatings (50 μm) will be applied to all piping to reduce permeation by a factor of 10 to 10,000

0 10 20 30 40 500.0

0.5

1.0

1.5

2.0

Triti

um p

ress

ure

abov

e Pb

-17L

i (Pa

)

Number of pulses

Expanded view

0 1000 2000 30000.0

0.5

1.0

1.5

2.0

Number of pulses

0 1000 2000 3000Number of pulses

0

100

200

300

400

Triti

um re

leas

e (m

g-T/

a) Helium piping

Pb-17Li piping

ITER limit

077-05/rs

TBM Meets ITER General Safety Requirements

• TBM pressurization of the VV, vaults and test cell is within ITER acceptance criterion

• TBM FW beryllium oxidation does not result in hydrogen generation that exceeds ITER limit of 2.5 kg

• TBM system Pb-17Li inventory is over 0.28 m3, but based on Pb-17Li/water reaction data the quantity of hydrogen generated will be less than the ITER limit of 2.5 kg

• Helium inventory is ~17 kg for the FW helium cooling loop, which is less than ITER limit of 40 kg

• TBM decay heat removal demonstrated

• Tritium permeation may require a permeation barrier for piping to meet ITER limits

077-05/rs

LM Pump

HX

TBM

TBM BypassValve

Cold PbLi Supply Pipe

Hot PbLi Return Pipe

Storage Tank

Tritium Extraction

Bypass control can be used to shunt coldPbLi away from the TBM operation, while themajority of the loop is operated at lowtemperature

Hot He to TCWS Building

Cold He from TCWS Building

Hot He to TCWS Building

Cold He from TCWS Building

Primary He flowrate andtemperature is alsocontrollable so thattemperature differencescan be studied even in lowtemperature experiments

PbLi flowrate and coldleg temperature arecontrollable so thatlow temperatureoperation is achievedand a wide variety ofexperimentalconditions can beproduced

Pb-17Li by-pass Loop Schematic(To control Pb-17Li temperature

077-05/rs

We followed closely the U.S. Position on PIE Needs

• A dedicated TBM hot cell is needed, but only limited PIE should be performed at the ITER site since each party already has extensive facilities for conducting specimen preparation, examination and testing.

• To avoid the expense and difficulty of shipping a large radioactive component the main function of the TBM hot cell should be to perform optical and other nondestructive examination, sectioningand preparation of samples for shipment, diagnostics instrument replacement, replacement of specimen cassettes, limited repairs,and perhaps simple mechanical testing of pre-machined specimens.

077-05/rs

TBM Hot Cell Requirements

• Capability to handle large activated metal/ceramic components, gas, and liquids.

• Remote visual equipment and camera to inspect and record condition of the TBM and guide cutting operations.

• Equipment for measuring dimensional changes.• Nondestructive inspection equipment such as UT and ET for

examination of large components.• Equipment for removal and temporary storage of specimen cassettes.• Cutting tools for sectioning the TBM into smaller, more manageable

pieces.• Capability to prepare pieces (and specimens) removed from a TBM

for shipment to the country of origin.• Simple (tensile, Charpy impact) mechanical testing of pre-machined

specimens over a range of temperatures and possibly under vacuum/inert gas.

077-05/rs

Key Tasks to be Addressed During Preliminary Design

Mechanical: More details including flow distributions and fabrication procedureNeutronics: More details and 3-D analysis including TBM shield designPbLi MHD design: More details for the whole loop, including natural convection

effectsHe thermal hydraulics: More details on first wall one-side roughening,

performance and flow distributions for the helium loopsTritium: Design of the tritium extraction and impurity control systemsDisruption analysis: Complete conceptual assessment and work on detailsSafety: Review and work with ITER, continue to provide safety design guidanceBy-pass loop: More design detailsPIE and hot-cell: Module needs to be designed to be compatible with PIE and hot-

cell requirementsAncillary equipment: More design and system detailsDiagnostics: Module design to accommodate measurement needs

077-05/rs

SUMMARY

• Using RAFS as the structural material, He as the FW/structure coolant, Pb-17Li as the self-cooled tritium breeding material and SiCf/SiC as FCI conceptual DCLL-TBM assessment was completed.

• For the reference reactor design, ηth>40% can be projected.• For the ITER TBM design, the following areas were evaluated: mechanical design, neutronics and

activation, He cooled FW/structure thermal hydraulic, Pb-17Li circuit including MHD effects with FCI, structural analysis, LOCA, disruption impact, by-pass Pb-17Li loop, tritium extraction, safety, and ancillary equipment on the transporter and TCWS, including PIE and hot cell requirements.

• SiCf /SiC composite FCI is the key for high thermal performance.• The DCLL TBM design can satisfy all ITER TBWG requirements.• A test plan for DCLL has been drafted.• The DCLL DDD report is to be finalized in August.• The DCLL TBM concept is ready to proceed to the preliminary design phase.

Summary